Long-term installations of the DC-potential drop method in four nuclear power plants and the accuracies thereby obtained for monitoring of crack initiation and crack growth

After a total monitored operational timescale of almost five years on long-term installations, both in the laboratory and in four nuclear power plants, evidence can be put forward that the DC-potential drop method is now, at its current stage of development, suitable for inspecting and monitoring material regions such as, e.g. weld seams in pipework, for crack initiation and crack growth at power plant temperatures. This function can be performed with reliability and high sensitivity. The inspection and monitoring of cracks on the internal surface of the pipework can also be carried out from the external surface. The studies have shown that the method is basically able to monitor the growth of cracks found at discontinuous intervals using permanently installed potential probes, i.e. from plant inspection to plant inspection, while a transition to continuous monitoring is possible at any time. Thus a measure of redundancy can be provided for conventional ultrasonic and radiographic inspection, in particular for difficult to check austenitic weld seams. The method can also be seen as an alternative to the conventional techniques. When necessary, the cracks found can be measured more accurately than was previously possible with conventional ultrasonic and radiographic inspections. The total exposure to radiation can be reduced in comparison to other methods of inspection.     

Waveguide ultrasonic liquid level transducer for nuclear power plant steam generator

This paper deals with the development and utilization of a new acoustic method to control the level of the coolant in industrial power-generating equipment. The development research was carried out on the basis of the method of acoustic impedance, based on measuring attenuation of a longitudinal ultrasonic wave in a waveguide sensing element. Ultrasonic attenuation is determined by the degree of submergence of the sensing element in liquid. For this purpose waveguide ultrasonic transducers are used, made with specially designed technology (waveguide acoustic transducers—WAT-technology). The transducers are adapted for operation under extreme conditions in the water coolant with temperatures up to 350°C and pressure up to 20 MPa. The paper tackles: principle of operation of transducers, methods of testing under laboratory conditions, design and results of calibration in a thermophysical test facility, results of the operational trial of the level transducer at the nuclear power plant in Zaporozhye.     

Development of very-high-density low-enriched-uranium fuels

Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm⁻³ U₃Si₂–Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8–9 g cm⁻³ in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work.     

Part-through cracks in pipes under cyclic bending

The propagation of a circumferential external surface crack in a metallic round pipe under cyclic bending loading is examined through a two-parameter theoretical model. A finite element analysis is carried out to determine the stress-intensity factor distribution along the front of the flaw, which is assumed to present an elliptical-arc shape with aspect ratio α=a_el/b_el (a_el, b_el=ellipse semi-axes). The relative depth ξ of the deepest point on the front is equal to the ratio between the maximum crack depth, a, and the pipe wall thickness, t. The parameter R/t, with R=internal radius of the pipe, is made to vary from 1 to 10. The fatigue growth of the surface flaw occurs according to preferred paths in the diagram of α against ξ.     

Analytical study on dynamic buckling of reactor containment vessel

The dynamic buckling of a reactor containment vessel under earthquake conditions is evaluated using a nonlinear finite element method. It is based on the four-node MITC (mixed interpolated tensorial components) shell element originally proposed by K.J. Bathe, which has been modified by the authors to include the effect of large rotational increments. At first, the buckling modes for a thin cylindrical shell under a simplified base excitation were classified, then the dynamic buckling analysis of a typical PWR steel containment vessel was carried out, considering both geometrical and material nonlinearities, to compare the results with those of a conventional static analysis. It was found that the global shear buckling of a steel containment vessel occurred under a load level several times greater than the design earthquake, and the buckling load estimated by the conventional analysis was smaller than the buckling load estimated by the dynamic analysis.     

Benchmark calculations for mechanical stresses upon a transport cask

In this paper, benchmark calculations are carried out using the FEM code ADINA. Three drop tests with a DCI cask carried out in Japan are investigated. The drop orientations comprise the 9 m drop with impact flat onto the cylindrical shell of the shock absorbers at the top and bottom part as well as the 1 m drop with impact of the bar in the center of the cylinder wall. Calculated results are compared to the experimental ones. It is found that the calculations meet the measurements conservatively.     

Experiment and numerical simulation of double-layered RC plates under impact loadings

This paper is to investigate and to propose a method to improve the impact resistance of reinforced concrete (RC) plates against projectile impact, and the damage of double-layered RC plates is examined experimentally and simulated analytically. In tests, a projectile launching apparatus, which is a 40 mm smooth-bore airgun, was used. Based on experimental results, numerical simulations with the DYNA-3D code, which takes account of the dynamic constitutive law of concrete, were done to find the applicability of the present computer code to the analysis of double-layered RC plates under high-speed impact loadings. In this study, the impact resistance of concrete plate is defined as the degree of local damage. In both experiments and numerical simulations, the effect of double-layering on the impact resistance is discussed.     

Cladding metallurgy and fracture behavior during reactivity-initiated accidents at high burnup

High-burnup fuel failure during a reactivity-initiated accident has been a subject of safety-related concern. Because of wide variations in cladding metallurgical and simulation test conditions, it has been difficult to understand the complex failure behavior observed in tests in the SPERT, NSRR and CABRI reactors. In this paper, we propose a failure model that is based on temperature-sensitive tensile properties and fracture toughness. The model assumes that dynamic fracture toughness and high-strain-rate tensile properties of high-burnup cladding are sensitive to temperature and exhibit ductile–brittle transition phenomena similar to those of bcc alloys. Significant effects of temperature and shape of the pulse are predicted when a simulated test is conducted near the cladding material's ductile–brittle transition temperature. Temperature dependence of tensile properties and fracture toughness is, in turn, sensitive to cladding microstructural characteristics such as density, distribution and orientation of hydrides; distribution of oxygen in the metallic phase; and irradiation-induced damage. Because all of these characteristics are strongly influenced by corrosion, the key parameters that influence susceptibility to failure are oxide layer thickness and hydriding behavior. Therefore, high-burnup fuel failure is predicted to be more sensitive to local cladding corrosion (e.g. grid span location) than to fuel burnup.     

Material properties for reactor pressure vessels and containment shells under dynamic loading

The effects of high strain rate, dynamic biaxial loading and deformation mode (tension, shear) on the mechanical properties of AISI 316 austenitic stainless steel in as-received and pre-damaged (creep, LCF) conditions are reported. This research was conducted to assess the performances of the containment shell of fast breeder reactors. The results of this research have been utilized to prepare similar investigations for SA 537 Class 1 ferritic steel used for the containment shell of LWR. The first results of these investigations are reported. A programme to study the mechanical properties of plain concrete with real size aggregate at high strain rate is described.     

Defect production considerations for gamma ray irradiation of reactor pressure vessel steels

Motivated by the recent interest in gamma ray embrittlement of nuclear reactor pressure vessels (RPVs), calculations were performed to evaluate aspects of defect production by gammas in iron and steel. In addition to determining displacement damage cross-sections, the atomic recoil energy dependence of gamma-induced defect production was described by integral recoil damage spectra, W(T), and their associated median recoil damage energies, T_1/2. These latter characterizations, should be particularly useful in evaluating the contribution of gamma ray generated defects to microstructural changes causing radiation embrittlement. The results for monoenergetic gammas, as well as for gamma rays with a spectrum of energies characteristic of a RPV, reveal T_1/2 values of <100 eV, about three orders of magnitude smaller than for fast-neutrons, the radiation of primary concern in previous embrittlement studies. The relative contributions of various gamma interactions to defect production, as well as the role of light alloy element-induced secondary displacement mechanisms, are also considered.     

Application of fracture mechanics in the design and operation of PWRs in France

Fracture mechanics techniques are used to evaluate the risks associated with material aging, the occurrence of fabrication defects or stress corrosion cracks, and to determine the margins with respect to fast fracture resulting from the design and manufacturing practices. Important programs are underway to improve the accuracy and domain of validity of the fracture mechanics methods.     

Isotope c(U)/c(U) relation in PIK reactor fuel elements

Isotope c(²³⁴U)/c(²³⁵U) relation in PIK reactor fresh fuel elements is investigated by non-destructible γ-spectroscopic method. The 60 PIK fuel elements were measured. Isotope c(²³⁴U)/c(²³⁵U) relation in PIK fuel elements equals (0.82±0.07)%.     

Alternative methods for postweld treatment of austenitic pipe welds to increase the operational safety of BWR plants

Experience has shown that austenitic piping is susceptible to IGSCC in the weld root area under BWR service conditions. Besides non-optimized materials, the residual stresses which are an inherent result of conventional welding processes are also responsible for this susceptibility. In the past, mechanical, thermal and welding post-treatment processes were developed and used with the objective of reducing tensile stresses in the root area. This paper discusses past experience and more recent developments, in particular the latest results with pipe welds treated by means of welding processes (last pass heat sink welding). These measures are suitable for producing compressive stresses in the medium-swept ID-HAZ of austenitic welds, or to at least significantly reduce the tensile stresses and thus practically eliminate the risk of IGSCC.     

Experimental assessment of neutron fluence for the Kozloduy Unit 1 pressure vessel by scraps and cavity dosimetry

In accordance with the Safety Improvement Program the neutron exposure onto the Kozloduy Unit 1 pressure vessel has been evaluated. The conservative fluence values have been used for the reembrittlement assessment of the pressure vessel. The verification of the calculated neutron fluence has been provided by comparing the calculated and experientially determined induced activities of the scraps taken out from the weld four and of the copper, iron, and niobium detectors irradiated in the cavity behind the vessel.     

Determination of more realistic Ke,r-factors for simplified elastic–plastic analysis

According to the relevant KTA-Rules, e.g. KTA 3201.2 (KTA, 1994), strain correction factors—K_e-factors—have to be used in the fatigue analysis of pressurised components if the strain intensity ranges are determined by elastic analyses, and if in this case the range of primary plus secondary stress intensity exceeds a certain limit. This limit is three times the design stress intensity value, S_m, and thus approximately corresponds to twice the value of the 0.2% strain limit. The relations given in the above-mentioned rules to determine the K_e-factors for considering plastification have proved to be very conservative in many cases compared with the strain intensity ranges that were determined by complete elastic–plastic analyses. In order to improve the validity of the fatigue analysis, the topic of `Performance of fundamental work to prepare concrete proposals for realistic K_e,r-factors (strain correction factors) to consider plastification at large strain amplitudes' was one of the subjects of the BMU project SR 2063. Work on this topic was jointly performed by GRS and Siemens. In summary, the result was that the proposed realistic K_e,r-factors present a real alternative to the K_e-factors of the regulations; the latter serve a mostly conservative registration of the observed elastic–plastic strain but cannot be explained in terms of physics and are not formulated in a manner adequately specific of any material. The exemplary verification calculations that have been performed so far show, furthermore, that the proposed realistic K_e,r-factors can be easily determined and also deliver sufficiently conservative results. This new method therefore has great potential which, however, still has to continue to be verified by further calculations before it can be included in the KTA-Rules.     

Application of Leak-Before-Break Concept in VVER 440-230: Results and consequences for TACIS-program 91/1.2

The project TACIS 91/1.2 `Primary Circuit Integrity, Application of Leak-Before-Break (LBB) Concept' was performed within the scope of the TACIS-program (Technical Assistance for the Community of Independent States) of the European Community. The application of the LBB concept was carried out for the primary piping (main coolant lines (MCL) and surgelines (SL)) of the first generation of the Russian VVER NPP's of the type VVER 440/230 of the nuclear power plants Kola 1 and 2 (K1, K2) and Novovoronesh 3 and 4 (NV3, NV4). The work was performed by the Consultant (Siemens, EdF, Framatome) and Russian subcontractors (headed by Vniiaes). The paper describes the comparison of the relevant safety concept, the scope of the LBB-program, the applied LBB-procedures (German and NUREG 1061 procedure), the results obtained and the recommendations (analysis of seismicity, if necessary LBB analysis taking into account additional optimized seismic supports and their installation for NV3 and 4 adaption of inservice inspection (ISI) and leak monitoring systems (LMS)). Proposals have been made for future work to improve the safety of these plants. This project was financed by the European Community's TACIS Programme, which provides grant finance for know-how to foster the development of marked economies and democratic societies in the New Independent States and Mongolia.     

A generalized model for radiation-induced amorphization and crystallization of U3Si and U3Si2 and recrystallization of UO2

A rate-theory model of radiation-induced amorphization and crystallization of U₃Si during ion irradiation has been generalized to include U₃Si₂ and UO₂. The generalized model has been applied to ion-irradiation and in-reactor experiments on U₃Si and U₃Si₂ and provides an interpretation for the amorphization curve (dose required to amorphize the material as a function of temperature), for the ion-radiation-induced nanoscale polycrystallization of these materials at temperatures above the critical temperature for amorphization, as well as for the role of the small crystallites in retarding amorphization. An alternative mechanism for the evolution of recrystallization nuclei is described for a model of irradiation-induced recrystallization of UO₂ wherein the stored energy in the UO₂ is concentrated in a network of sinklike nuclei that diminish with dose due to interaction with radiation-produced defects. The sinklike nuclei are identified as cellular dislocation structures that evolve relatively early in the irradiation period. The complicated kinetics involved in the formation of a cellular dislocation network are approximated by the formation and growth of subgrains due to the interaction of shock waves produced by fission-induced damage to the UO₂.